ML19290C375
| ML19290C375 | |
| Person / Time | |
|---|---|
| Site: | 05000502, 05000503 |
| Issue date: | 04/30/1976 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0051, NUREG-0051-01, NUREG-51, NUREG-51-1, NUREG-75-092, NUDOCS 8001100803 | |
| Download: ML19290C375 (43) | |
Text
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t provide two licensed senior reactor operators and three licensed reactor operators per operating shift for two-unit operation. Further, we require that the applicant commit to this requirement in the application to assure that an acceptable number of trained personnel will be available to operate the facility upon completion.
This matter still is under discussion between the staff and the applicant and we hope to be able to report its resolution when we meet with the Advisory Comittee on Reactor Safeguards. For now, however, it remains unresolved.
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14.0 INITIAL TESTS AND OPERATIONS This section of the Safety Evaluation Report indicated that planned preoperational testing was not entirely in accordance with methods described in Regulatory Guide 1.79.
We stated that we would require the applicant to develop acceptable testing methods during the Operating License review.
As reported in Section 6.3.4 of this Supplement, the applicant, in Amendment 11 to the Preliminary Safety Analysis Report, committed (1) to conduct the low pressure safety injection recirculation test taking suction from the containment sump and, (2) to conduct the low pressure and high pressure safety injection flow tests at anbient conditions, all in accordance with applicable portions of Regulatory Guide 1.79.
The staff has reviewed these two commitments and has determined that they are acceptable.
We therefoae conclude that the initial test program will use acceptable testing methods and this matter is resolved.
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15.0 ACCIDENT ANALYSES 15.3.2 Feedwater System Piping Breaks The Safety Evaluation Report indicates t%t there are unresolved questions regarding the calculational methods used in the analysis of the effects of feedwater system piping breaks. The staff is working with the reactor vendor on a generic basis to resolve these questions. The applicant has agreed to accept the results of this generic review effort. We find this to be acceptable at this stage of the licensing process.
15.3.3 Spectrum of Steam Pipinq Failures inside and Outside of Containment The Safety Evaluation Report states that the staff has accepted the predicted consequences of a steam line break accident, but that there still are questions regarding the calculational methois used. The staff is working directly with the reactor vendor on this matter. The applicant has agreed to accept the results of this generic review and we find this commitment to be acceptable at this stage of the licensing process.
15.3.4 Spectrum of Piping Breaks Within the Reactor Coolant Pressure Boundary 15.3.4.1 Introduction The applicant submitted an evaluation of the Emergency Core Cooling System performance in Amendment 10 on October 2, 1975 (small break Loss of Coolant Accident) and in Amendment 12 on February 6,1976 (major reactor coolant system pipe ruptures) pursuant to the requirements of the Comission's regulations 10 CFR 50.46. The analyses submitted were based on the approved Westinghouse Emergency Core Cooling System evaluation model.
15.3.4.2 Emergency Core Cooling System Analysis The applicant's submittal of a large break Loss of Coolant 1ccident analysis was limited to a spectrum of three guillotine breaks, which was specific for the Koshkonong Nuclear Plant, Units 1 and 2.
To supplement the analysis of the three breaks, tne applicant referenced WCAP-8356, Westinghouse ECCS Plant Sensitivity Studies; WCAP-8472, the Westinghouse ECCS Evaluation Model - Supplementary Information; and WCAP-8573 Westinghouse ECCS - Th*ee Loop Plant 17 x 17 Sensitivity Studies; which demon-strated that the guillotine breaks are the worst cases for this plant type.
The analyses submitted identified the worst break size as the double-ended cold leg guillotine break with a Moody multiplier of 0.4.
The calculated peak clad temperature was 2146*F, within the acceptable limit of 2200'F (as specified in 10 CFR 50.46(b)).
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In addition, the maximum local metal / water reaction of 7 7% and a total core-wide metal / water reaction of less than 0.3% were well below the allowable limits of 17%
and 1%, respectively. The analyses were performed based on an assumed total peaking factor of 2.18 at 102% of rated Nuclear Steam Supply System power level of 2910 megawatts-thermal, with a peak linear power density of 11.85 kilowatts per foot.
The Emergency Core Cooling System analysis indicates that the maximum power density in the core at full power must not exceed that associated with a peaking factor of 2.18.
The applicant, however, has not shown how the power distribution will be unaged so that at full power the maximum peaking factor will not exceed 2.18.
Heretofore, in the absence of an appropriate plant-specific analysis, we have only credited Constant Axial Offset Control, which the applicant proposes to employ, with the capability to limit peaking factors to less than 2.32.
Therefore, we require that the applicant either present an analysis for Koshkonong that demonstrates that Constant Axial Offset Control alone can limit peaking factors to less than 2.18 or commit to install if necessary, an Axial Power Distribution Monitoring System which in concert with Constant Axial Offset Control has been shown capable of assuring that the peaking factor will not exceed limits as low as approximately 2.
The applicant has committed to install an Axial Power Distribution Monitoring System, if necessary.
This commitment is acceptable to the staff.
The small break analysis which was submitted in Amendment 10 to the Preliminary Safety Analysis Report on October 2, 1975, included a three-break spectrum and referenced WCAP-8356, a generic Westinghouse topical report which documented additional break analyses. The small break analysis, which identified the 3-inch pipe break as the limiting small break with a peak clad temperature of 1935'F, demonstrates that the small break Loss of Coolant Accident is not limiting.
15.3.4.3 Emergency Core Cooling System Containment Pressure Evaluation The Emergency Core Cooling System containment pressure calculations for the Koshkonong Nuclear Plant. Units 1 and 2 were done using the Westinghouse Emergency Core Cooling System evaluation mode. The Uclear Regulatory Comission staff has reviewed Westinghouse's containment inodel and found it acceptable for Energency Core Cooling System evaluation. The staff required, however, that justification of the plant-dependent input parameters used in the analysis of each plant be submitted for our review. This information was submitted in Amendment 12 to the Preliminary Safety Analysis Report. The staff has reviewed this infonnation and based on confirmatory analyses, has concluded that the Emergency Core Cooling System containment pressure analysis is reasonably conservative. Therefore, the calculated containment pressures are in accordance with Appendix K to 10 CFR Part 50 of the Commission's regulations and are acceptable.
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15.3.4.4 Single Failure Criterion Appendix K to 10 CFR Part 50 of the Commission's regulations requires that the combir.ation of Emergency Core Cooling System subsystems to be assumed operative shall be those available after the most dameging single failure of the Emergency Core Cooling System equipment has occurred. The worst single failure which could minimize the emergency cooling available to cool the core anj to provide maximum containment cooling was identified by Westinghouse as the loss of a low pressure Emergency Core Cooling System pump. As stated in a letter from D. B. Vassallo to C. Eicheldinger, dated May 30, 1975, subject: "NRC Staff Review of the Westinghouse Emergency Core Cooling System Evaluation Model," the staff concluded that the application of the single failure criterion was to be confirmed during subsequent reviews of Westinghouse designs for conformance to the criteria of Appendix K.
A review of the piping and instrumentation diagrams for the Koshkonong units indicated that the inadvertent actuation of specific motor-operated valves could affect the appropriate single failure assumptions. Toe staff previously had identified six such motor-operated valves and had indicated that removal of electrical power to the valve operators would satisfy the single failure criterion, lne following is a complete list of the valves identified by the staff that will require modification:
Motor Operated Valve Number Component Function Failure Mode 8808A, B, & C Accumulator Isolation valves Inadvertent closing of these valves would stop accumulator flow 8889 Isolates the Residual Heat Inadvertent opening of this Removal system from the valve would permit injection Reactor Coolant System into Reactor Coolant System hot leg hot leg and cause steam binding 8884, 8886 Isolates hot leg injection Opening of this valve during lines Emergency Core Cooling injec-tion and core reflood will allow injection into Reactor Coolant System hot leg and cause steam binding The staff has reviewed the consequences of failures of the above valves and has concluded that the following modifications are acceptable for complying with the single failure criterion:
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(1) During power operation, alternating current power will be disconnected to valves 8808A, B and C by racking out of breaker at the motor control center.
These valves will be locked in the open position, since inadvertent actuation of any of these valves would stop accumulator flow.
(2) During power operation, alternating current power will be removed from valves 8884, 8886, and 8889. These valves will be in their closed position and the design will provide for power to be restored from the control room for switchover to recirculation mode operation. Necessity for this action is explained in the above table and in U.S. Nuclear Regulatory Commission Standard Review Plan, Office of Nuclear Reactor Regulation, Appendix 7A, Branch Technical Position E!CSB-18, Application of the Single Failure Criterion to Manually Controlled Electrically Operated Valves.
15.3.4.5 Long-Te., Boron Concentration Buildup The applicant has proposed to provide long-term core cooling by changing over from cold leg to hot leg recirculation at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a Loss of Coolant Accident.
The Nuclear Regulatory Coninission staff has reviewed the Emergency Core Cooling System fcr Koshkonong Units 1 and 2 and has found that the system design will permit this implementation procedure. This emergency operating procedure will be reviewed by the staff when it is made available during the evaluation of the Final Safety Analysis Report.
15.3.4.6 Submerged Valves The applicant has identified three motor-operated valves that will be submerged following a Loss of Coolant Accident. Those valves whose motor operators are locr.ted below the flooded elevation are listed below.
Motor Operated Valve Number Mot 3r Bottom Elevation Description 8808A, B. & C 49' 7" Accumulator tank isolation valves These valves are required to be functional following a LOCA. They are normally open, but also receive a Safety Injection Signal in the event of a Loss of Coolant Accident.
The accumulator blowdown is completed within the first few minutes after the accident and well before the motor operators could become submerged. Therefore, the applicant submitted that subsequent flooding has no adverse effects and the staff agrees that the design is acceptable.
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15.3.4.7 Conclusions Based on our review, the staff concludes that:
(1) The Loss-of-Coolant Accident analyses that were performed are wholly in accordance with the requirements of Appendix K to 10 CFR 50.
(2) The Emergency Core Cooling System cooling performance conforms to the peak clad temperature and maximum oxidatien and hydrogen generation criteria of 10 CFR 50.46.
(3) The performance of the Emergency Core Cooling System will be adequate despite any postulated failure of a single active component.
(4) Adequate systems are available to provide long-term core cooling to the reactor vessel.
15.4 Anticipated Transients Without Scram The Nuclear Regulatory Comission's staff issued its status report on anticipated transients without scram for Wastinghouse reactors on December 9,1975. The Advisory Committee on Reactor Safeguards reviewed this report, and the report on other reactor vendors, at its 189th meeting on January 8-10, 1976. The Committee interim report on this matter is contained in its letter dated January 14, 1976. This letter is attached as Appendix D to this Supplement.
In its status report, the Comission's staff has identified certain outstanding issues which must be resolved to obtain an approved analytical model for plants. We expect this to be completed by June of this year. We currently are developing a program for implementation of our requirements by applicants and licensees. When this program has been completed and the required information has been received, we will issue a supplement to our status report.
In its interim report, the Advisory Comittee on Reactor Safeguards stated that it endorses the general approach and safety objectives adopted by the Nuclear Regulatory Comission's staff. The Comittee expects to complete its review of anticipated transients without scram after further information has been developed and the Commission's staff has completed its evaluation.
The applicant has advised the staff that it does not agree with the staff position on this matter and that it supports the Westinghouse position as stated in a letter from C. Eiche1dinger to R. Heineman dated February 2, 1976. Nonetheless, the staff intends to require implemntation of its position for the Koshkonong plant at such time as we complete development of the implementation program.
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17.0 QUALITY ASSURANCE 17.5 Implemen ta tion The ongoing quality assurance program is acceptable to the staff. However, certain activities such as prcparation of procurement documents and long lead procurement of equipment are not now underway due to restrictions imposed by the Public Service Commission of Wisconsin on fund expenditures. Audits of these types of activities normally are performed by the Office of Inspection and Enforcement prior to granting of a construction permit. Such audits will be r.ade when the activities are initiated.
Meanwhile, inspections and audits performed to date have indicated no reason to suspect that the applicant's quality assurance program will not be acceptable.
A corporate reorganization effective April 1,1976, places the quality assurance activities under control of the Senior Vice President of Wisconsin Electric Power Company, rather than under the Executive Vice President as had previously been the The applicant has stated that this change will have minimum effect on quality case.
assurance operations. Nonetheless, this matter will be reviewed further to assure that the quality assurance program still is acceptable. It is considered to be unresolved pending this further review.
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18.0 REVIEW BY THE I.DVISORY COMMITTEE ON REACTOR SAFEGUARDS The application was considered by a Subconnittee of the Advisory Committee on Reactor Safeguards on October 17, 1975, in Fort Atkinson, Wisconsin, and on December 3, 1975, in Washington, D.C.
The staff and the applicant met with the full Committee on December 5, 1975, during its 188th meeting. As a result of these meetings, the Advisory Committee on Reactor Safeguards issued an " Interim Report on Koshkonong Nuclear Plant, Units 1 and 2" on January 15, 1976. This report is appended as Appendix C.
The staff and the applicant will meet again with both the Subcommittee and the full Connittee to discuss those matters identified by the Committee in the interim report.
These meetings and others, if necer.sary, will result in another report by the Committee which we will address in a later supplement to the Safety Evaluation Report. The current status of the matters mentioned in the interim report is presented in the following paragraphs.
The Coculttee reported that It was continuing its review of the Emergency Core Cooling System analysis submitted by the applicant, in view of a revised evaluation model which the applicant intended to use for evaluation of the Wisconsin Utilities Project nuclear steam supply systems. This revised analysis was submitted in Amendment 12 to the Preliminary Safety Analysis Report. The results of the staff review of this revised analysis are presented in Sections 6.2.7 and 15.3.4 of this supplement.
The Committee also indicated its desire to consider further the matter of design values for horizontal acceleration to be used for the Safe Shutdown Earthquake and the Operating Basis Earthquake. The results of the additonal staff review are presented in Section 2.5.2 of this supplement.
The Committee recommended that the design of the Wisconsin Utilities Project units incorporate a requirement for a loose parts monitor. The staff has considered this recommendation and advised the applicant that a loose parts monitor would be required on each of the units. The applicant now has committed to provide a loose parts monitor and this matter is, therefore, resolved.
The Conpittee indicated its desire to be kept informed of the status of the staff review of the applicant's analysis of Anticipated Transients Without Scram and the capability of the liquid and gaseous radwaste systems to meet the design objectives of Appendix I to 10 CFR Part 50. The current status of the Appendix I evaluation is presented in Section 11.1 of this supplement. The status of the staff review of Anticipated Transients Without Scram is presented in Section 15.4 of this supplement.
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The Committee recommended that the staff and the applicant review the Koshkonong plant for design features that could significantly reduce the possibility and conse-quences of sabotage. As stated in Section 13.6 of the Safety Evaluation Report, the applicant's plans are in conformance with Regulatory Guide 1.17, which is appropriate for this stage of the review. The staff considers that design featun s provided in the Koshkonong Nuclear Plant, Units 1 and 2, that mitigate the consequences of accidents also substantially reduce the chance that an act of sabotage could result in jeopardizing the public health and safety. At the operating license stage, the staff will review the plan for the protection of the Koshkonong plant against acts of industrial sabotage. In addition, the staff will continue in its generic effort regarding nuclear power plant design features that could reduce the vulnerability to and consequences of sabotage.
Finally, the Comittee requested to be kept informed of the staff and applicant review of design features intended to prevent the occurrence of damaging fires and to minimize the consequences to safety-related equipment should a fire occur.
This subject area has been under intensive review by the staff for the past year as a result of the fire that occurred at the Browns Ferry Plant on March 22, 1975. A Special Review Group was established shortly after that fire to identify the lessons learned from the fire and to make recommendations as appropriate. The efforts of this group are reported in NUREG-0050, "Recomendations Related to Browns Ferry Fire," published in February 1976. Concurrently, the staff has been reviewing its guidelines for fire protection. These efforts have resulted in proposed revisions to the staff position on fire protection which, when approved, will become the basis for further review of fire protection for the Koshkonong units as well as for other plants.
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APPEf; DIX A CHRONOLOGY OF RADIOLOGICAL REVIEW (Continued)
October 14, 1975 Amendments #10 to WUP PSAR and Koshkonong Site Addendum received October 17, 1975 Safety Evaluation Report issued October 17, 1975 Meeting with ACRS Subcommittee; Subcommittee site visit November 4, 1975 Meeting with applicant to discuss open items remaining from radiological safety review November 5,1975 Initial appeal meeting with applicant for discussion of remaining open items November 21, 1975 Meeting with applicant to discuss revised ECCS model for WUP plants November 21, 1975 Appeal meeting with applicant November 21, 1975 Amendments #11 to WUP PSAR and Koshkonong Site Addendum received December 3, 1975 Meeting with ACRS Subcommittee December 5, 1975 Meeting with ACRS January 15, 1976 ACRS interim report issued February 17, 1976 Amendments #12 to WUP PSAR and Koshkonong Site Addendum received March 3, 1976 Letter to applicant with decisions on appeal itens March 10, 1976 Letter to applicant requesting statement of position on unresolved issues March 31, 1976 Letter from applicant stating position on unresolved issues A-1
APPENDIX C ADVISORY COMMITTEE ON REACTOR SAFEGUARDS NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 January 15, 1976 lbnorable William A. Anders Chairman U. S. Nuclear Regulatory Ormission Wahington, DC 20555 SUR7EX:T: Dm: RIM REPORP W KOSHKCNCNG NUCLEAR PIRTf, UNITS 1 & 2
Dear Mr. Anders:
During its 189th meeting, January 8-10, 1976, the Advisory Comittee on Reactor Safeguards empleted a partial review of the application of the Wisconsin Electric Ibwer Conpany, Wisconsin Ibwer and Light Coupany, Wisconsin Public Service Corporation, and Madison Gas and Electric Coupany (Applicant) for a permit to construct the Koshkonong Nuclear Plant, thits 1 & 2.
%is project had been previously considered at the Comittee's 188th meeting and at Subcomittee 1c
'ngs in Ft. Atkinson, Wisconsin, on October 17, 1975 and Washington on Decm ber 3, 1975. Mernbers of the Comittee visited the site o.. October 17, 1975. During its review, the Comittee had the benefit of discussions with representatives and consultants of the Applicant, Westinghouse Electric Corporation, Stone and Webster Corp-oration, and the Nuclear Regulatory Comission (NRC) Staff. We Comittee also had the benefit of the references listed.
We application to build the the Koshkonong Nuclect Plant is a part of the application, designated the Wisconsin Utilities Project (WUP), for licenses to construct and operate one or nore standardized nuclear power plants at one or acre sites in the State of Wisconsin, using the duplicate plant option, Appendix N to 10 CFR Part 50. he site-related aspects specific to the Koshkonong plant are contained in a Site Addendum to the WUP Preliminary Safety Analysis Report.
He Koshkonong plant will be located on an 1109 acre site in Jefferson County, Wisconsin,11 miles northwest of Janesville, the nearest population center (1970 population 46,246). % e minimum exclusion distance is 954 meters and the low population zone radius is three miles.
Each unit will utilize a 3-loop Westinghouse pressurized water reactor with 17x17 fuel assmblies to be operated at power levels up to 2775 MW(t).
%e nuclear steam supply system is similar in design to Virgil C. Sumer, thit 1 reported on by the Comittee in its letter of November 15, 1972.
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ibnorable William A. Anders January 15, 1976 During October 1975 Westinghouse sutznitted a revised EOCS evaluation model for review by the NRC Staff. %is review is nearing cortpletion. @e Applicant intends to use the approved revised zrodel for ECCS evaluation of the WUP nuclear steam systems. % e calculated reflooding rates and low peaking factor are of particular interest to the Comittee. We Comittee will continue its review of the WUP ECCS evaluation until the matter is resolved in a manner satisfactory to both the Comittee and the NRC Staff.
We Applicant and the NBC Staff both selected the tectonic province approach permitted by Appendix A to 10 CFR Part 100 to establish con-servative design values for horizontal ground acceleration. %e NRC Staff considered the applicable province to include Anna, Ohio, the site of a 1937 earthquake of intensity VII-VIII (MM). On this basis the NRC Staff believes that the design value for horizontal acceleration for the SSE should be 0.20g and for the OBE 0.103. %e Applicant is now examining proprietary data from oil exploration drilling in the Anna area, which he believes will show that the Arma earthquake was not a random earthquake but rather was associated with a local active fault. %e Applicant is also proposing decoupling of the OBE from the SSE. %is matter should be resolved to the satisfaction of the Comittee and the NRC Staff.
%e NRC Staff has not yet completed its review of: (1) the Applicant's analysis of Anticipated Transients Without Scram; and (2) the capability of the liquid and gaseous radwaste systens to meet the design objectives of Appendix I to 10 CFR Part 50. %e Comittee wishes to be kept informed.
Recent standardized safety designs for nuclear steam systems have included loose parts monitors. We Comittee recomends that a similar requirement be made a part of the WUP safety design. %e Comittee wishes to be kept informed.
We Comittee believes that the Applicant and the NRC Staff should review the Koshkonong Plant for design features that could significantly reduce the possibility and consequences of sabotage, and that such features should be incorporated into the plant design where practicable. %e Comittee wishes to be kept informed.
We Comittee recomends that the NRC Staff and the Applicant review the design features that are intended to prevent the occurrence of damaging fires and to minimize the consequences to safety-related equignent should a fire occur. We Comittee wishes to be kept infonned.
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Honorable William A. Anders January 15, 1976 Generic problems relatire to large water reactors are discussed in the Ox:nittee's report of March 12, 1975. Wese problems should be dealt with appropriately by the NBC Staff and the Applicant.
%e Cmunittee will emplete its review of this applicaticn whm the necessary additional information has been developed.
Sincerely, i
Dade W. Moeller 01 airman References 1.
Koshkonong Nuclear Plant thits 1 and 2, Preliminary Safety Analysis Report (August 1974) with Amendments 1 through 10.
2.
Koshkonong Nuclear Plant PSAR Site Addendum (Allgust.1974) with Amendments 1 through 10.
3.
Safety Evaluation Report NUREG-75/092 related to construction of the Koshkonong Nuclear Plant thits 1 and 2, October 1975.
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APPENDIX D ADVISORY COMMITTEE ON REACTOR SAFEGUARDS NUCLEAR REGULATORY COMMISSION WASHING TON. D. C. 20555 January 14, 1976 Honorable William A. Anders Chairraan U. S. I:uclear Pogulatory Cer aission 1:ashingtca, DC 20555 SUIk7ECT:
IITTERIM PSICIC Ctl M?rICIPATED TPAUSIDTIS iTI11IOUP SCRN! (AES)
Dear tir. Anders:
At its 109th meeting, January 8-10, 1976, the Advisory Cc:mittee on Practor Safeguards reviewed the 1:ucicar Pegulatory Comission Staff *n status reports on N ES in water reactors and the analyses and proposa e M four reactor vendors, '1he Babcock and Uilcox Company, Co:rbustion Engineering, Inc.,
V:estinghouse Electric Corporation, and General Electric Ccrpany, on this matter.
Subcomittee r,cetings were held with representatives of the vendors, and with ITC Staff in 1:ashington, DC, on D2ccaber 11-12, 1975, and on January 7, 1976. The Corraittee had the benefit of the docu:aents listcd.
7he Co.i:nittee cm:~2nted on the NPC Staff's proposal to revise the criteria for "C1 css A" plants, as catcgorized in 1: ASH-1270 "Inticipated Transients Without Scram for hater Cooled Ibuer Peactors," in a letter to Mr. Lee V.
Cossick dated Ct:tober 17, 1975. 'Ihe Cemittee had previously co:mr2nted on the regulatory position eventually published in 12SH-1270, in letters to Mr. L. l'anning 1:untzing dated April 16, 1973, and May 10, 1972.
'Ihe ACRS endorses the general aoproach and safety objectives adopted by the NEC Staff including the use of a goal of 10 7 per reactor year as the ra3xi-mum probability, from all causes, of AIUS with unacceptable consequences.
Implicit in the use of a probabilistic goal is the application of proba-bilistic nyathods in the analysis of the reactor systems. Although v2SH-1400 provides assistance in this area, data for some systems are still suffi-ciently sparce that engineering judgnant must be used, both in synthesizing the analytical models and in choosing appropriate input data.
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Ibnorable William A. Anders January 14, 1976 circumstances there is a need for some conservatism in the choice of models and the selection of data.
Even so, there are a number of approaches to the redelling that may prove to have equal validity and the ACPS stygests that both the ImC Staff and the vendors give further consideration to various alternatives.
For example it may be feasible to treat the *. fun-varying moderator terrporature coefficient probabilistically.
The Conntittee also recormnnds that vendors be encouraged to continue to make design changes that decrease the probability of transients that are likely to cause difficulty and to make improverrents that irreliorate the consequences. As appropriate these should be taken into account in the A'n:S analysis. Continuing attention should be given to improving the reliability of the reactor shutdo.m systems.
During the cource of these treetings cc:ments store made which indicated that in some cases the imC Staff necdcd further information from vendors in order to conclude its review of NIUS. The Cc:rmittee urges that agro-priate action be taken to obtain this information as soon as feasible.
'lho ACPS expects to co:mlete its review of nit:S after further information has been developed and the Staff has co:mleted its evaluation. The Committee urges that the imC Staff and the vendors expedite efforts in this regard.
Sincerely yours, N(
^'
Dade W.11oeller Chairman D-2
Ilonorable William A. Anders January 14, 1976 REFEPINCES:
1.
Intter dated January 5,1976, from ibbias W. T. Burnett, to T. G. ItCreless concerning request for oral presentation to ACRS tbrking Group on A1WS on January 7,1976 and to ACRS on January 8, 1976.
2.
Draft Copy of Presentation to the ACPS entitled, "The Atfr Issues: The Industry Consensus As "b Their Resolution," by T. W. T. Burnett, Clairman, ANS-51 AEff Working Group.
3.
NRC Staff's Status Report on Anticipated Transients Without Scram for Babcock and Ullcox Reactors, dated rnecaber 9,1975.
4.
NRC Staff's Status Report on Anticipated Transients Without Scram for Combustion Engineering, dated Deccaber 9, 1975.
5.
URC Staff's Status Report on Anticipated Transients Without Scram for General Electric, datd Eccerber 9,1975.
6.
Imc Staff's Status Report on Anticioated Transients Without Scram for Westirchouse, dated Decerber 9,1975..
7.
Proposed American National Standard Evaluation of Anticipated Transients Without Trip on Pressurized Water Peactor Plants, N661,Ibrch 1975.
8.
U.S. Atomic Energy Commission Pogulatory Staff's Technical Report on Anticipated Transients Without Scram For Water-Cooled rower Reactors, hASIl-1270, dated Septe:rber,1973.
D."